Integrated in-vessel neutron shield

ABSTRACT

To reduce size and mass of a nuclear reactor system, an integrated in-vessel shield separates the role of a neutron reflector and a neutron shield. Nuclear reactor system includes a pressure vessel including an interior wall and a nuclear reactor core located within the interior wall of the pressure vessel. Nuclear reactor core includes a plurality of fuel elements and at least one moderator element. Nuclear reactor system includes a reflector located inside the pressure vessel that includes a plurality of reflector blocks laterally surrounding the plurality of fuel elements and the at least one moderator element. Nuclear reactor system includes the in-vessel shield located on the interior wall of the pressure vessel to surround the reflector blocks. In-vessel shield is formed of two or more neutron absorbing materials. The two more neutron absorbing materials include a near black neutron absorbing material and a gray neutron absorbing material.

CROSS-REFERENCE TO RELATED APPLICATIONS

This application claims priority to U.S. Provisional Pat. ApplicationNo. 62/910,561, filed on Oct. 4, 2019, titled “Nuclear System for PowerProduction in Space,” the entirety of which is incorporated by referenceherein.

This application relates to International Application No.PCT/US2020/XXXXXX, filed on Oct. 4, 2020, titled “Nuclear Reactor CoreArchitecture with Enhanced Heat Transfer and Safety,” the entirety ofwhich is incorporated by reference herein. This application also relatesto International Application No. PCT/US2020/XXXXXX, filed on Oct. 4,2020, titled “Automatic Shutdown Controller for Nuclear Reactor Systemwith Control Drums,” the entirety of which is incorporated by referenceherein.

TECHNICAL FIELD

The present subject matter relates to examples of nuclear reactorsystems and nuclear reactors for power production and propulsion, e.g.,in remote regions, such as outer space. The present subject matter alsoencompasses a nuclear reactor that includes an in-vessel neutron shieldand a method for fabricating the in-vessel neutron shield.

BACKGROUND

Nuclear fission reactors include thermal or fast type reactors.Currently, almost all operating nuclear fission reactors are thermal.Nuclear fission reactors include nuclear fuel inside a nuclear reactorcore and a moderator to slow down fast neutrons so that nuclear fissioncan continue. Typically, the nuclear fuel is formed in cylindricalshaped fuel compacts or pellets. The fuel compacts are loaded into fuelpins or rods, cladded, and stacked inside the numerous columns of fuelelements in the nuclear reactor core.

The nuclear reactor core increases neutron fluence or neutron flux,which is the distance free neutrons move in a volume per time. Neutronfluence is necessary in nuclear reactors to generate heat and energy,but neutron fluence, particularly fast neutron fluence, or the neutronfluence of neutrons moving at a speed of 14,000 km/s or higher, is verydangerous to both matter and human life. Therefore, the high fastneutron fluence is decreased to safe levels at distances further fromthe nuclear reactor core, ideally to safe values just beyond the nuclearreactor itself. Some components of a nuclear reactor, such as neutronreflectors, perform this action: reflectors direct free neutrons backtoward the nuclear reactor core, thereby increasing neutron fluenceinside the nuclear reactor, improving energy efficiency, and decreasingneutron fluence outside the nuclear reactor, making the surrounding areasafer.

When building a conventional nuclear reactor system for conventionalterrestrial land applications, e.g., a nuclear power plant, the size(e.g., space or volume) and mass of the nuclear reactor is not a majorconcern. Typically, the actual thickness of the neutron reflector toreflect neutrons back to the nuclear reactor core for energy efficiencyis much greater than the required thickness needed to appropriatelyreduce fast neutron fluence to safe levels beyond the nuclear reactor.Therefore, the conventional nuclear reactor tends to have a very thickneutron reflector in order to reduce neutron fluence external to thenuclear reactor. Typically, the neutron reflector is much thicker thanrequired for optimal neutronic performance needed to reflect the freeneutrons back to the nuclear reactor core. Essentially, the neutronreflector additionally acts as a neutron shield.

However, the size and mass of a nuclear reactor system are importantfactors for nuclear thermal propulsion (NTP) and providing nuclear power(e.g., thermal and/or electrical power) in remote region applicationsincluding outer space, celestial bodies, planetary bodies, and remotesregions on Earth. For example, the mass of the nuclear reactor systemdirectly affects performance, such as power per mass, in the NTP andspace applications; and may add unnecessary manufacturing cost.Moreover, in the remote region applications, a smaller (more compact)form factor reduces construction costs and increases transportabilitywhile allowing operation of the nuclear reactor at high power densities.Accordingly, improvements to the conventional implementation of aneutron reflector as a space-inefficient neutron shield are needed.

SUMMARY

The various examples disclosed herein relate to nuclear technologies fornuclear reactor systems both for space or terrestrial land applications.The nuclear reactor system 100 advantageously reduces the size and massof the nuclear reactor 107 by implementing integrated in-vessel shield105 to separate the role of neutron reflector and the neutron shield andstill obtain optimal neutronic performance. The integrated in-vesselshield 105 brings radiation shielding into the pressure vessel 160 ofthe nuclear reactor 107, which allows for reduction of mass of thenuclear reactor system 100 and achieves a space and mass efficientneutron shield.

Integrated in-vessel shield 105 has the following benefits. First,in-vessel shield 105 reduces the required distance between the pressurevessel 160 and the active nuclear reactor core 101 and thus the totalsize of the pressure vessel 160. For example, in-vessel shield 105enables a thinner reflector 140 and smaller overall pressure vessel 160.Prior neutron reflectors unnecessarily increase diameter of the pressurevessel 160, leading to a larger diameter of the pressure vessel 160 thatis more expensive to manufacture and difficult to transport and field.Second, in-vessel shield 105 extends the life of the pressure vessel 160by reducing fast flux to the pressure vessel 160. Third, in-vesselshield 105 reduces activation outside of the nuclear reactor core 101and therefore makes the nuclear reactor 107 easier to install. Fourth,in-vessel shield 105 can be incorporated (e.g., retrofitted) into analready built nuclear reactor 107 or the design of a new nuclear reactor107.

To achieve breakthrough performance, in-vessel shield 105 includes anear black neutron absorbing material and a gray neutron absorbingmaterial to survive the radiation environment and thermal environmentthat is found inside of the pressure vessel 160 of the nuclear reactor107. For fabrication of the in-vessel shield 105, advanced 3D printingand spark plasma sintering methods can be employed. In-vessel shield 105is shaped as a plurality of in-vessel shield tiles 131A-N that shape thenear black neutron absorbing material and the gray neutron absorbingmaterial into an interlocking geometry pattern well-suited for placementon the interior wall 133 inside of the pressure vessel 160 and removesneutron streaming paths. In-vessel shield 105 solves the problem oflimiting fast fluence to the pressure vessel 160 in a volume-constrainednuclear reactor 107 and decreases fast flux to the pressure vessel 160with the outside. In-vessel shield 105 reduces the radiation fieldaround the nuclear reactor 107, which in turn decreases the activationin the area around the nuclear reactor 107 and facilitates installation.

An example nuclear reactor system 100 includes a pressure vessel 160including an interior wall 133 and a nuclear reactor core 101 locatedwithin the interior wall 133 of the pressure vessel 160. Nuclear reactorcore 101 includes a plurality of fuel elements 104A-N and at least onemoderator element 103. Nuclear reactor system 100 includes a reflector140 located inside the pressure vessel that includes a plurality ofreflector blocks 141A-N laterally surrounding the plurality of fuelelements 104A-N and the at least one moderator element 103. Nuclearreactor system 100 includes an in-vessel shield 105 located on theinterior wall 133 of the pressure vessel 160 to surround the reflectorblocks 141A-N. In-vessel shield 105 is formed of two or more neutronabsorbing materials. The two more neutron absorbing materials include anear black neutron absorbing material and a gray neutron absorbingmaterial.

An example method includes selecting two or more neutron absorbingmaterials to form an in-vessel shield 105 (block 610). The two or moreneutron absorbing materials include a near black neutron absorbingmaterial and a gray neutron absorbing material. The method furtherincludes eutectic sintering the near black neutron absorbing material tofabricate a ceramic absorbing powder (block 615). The method furtherinclude kinetically mixing the gray neutron absorbing material and theceramic absorbing powder to create an in-vessel shield mixture (block620). The method further includes cold press sintering the in-vesselshield mixture into an in-vessel shield 105 (block 625).

Additional objects, advantages and novel features of the examples willbe set forth in part in the description which follows, and in part willbecome apparent to those skilled in the art upon examination of thefollowing and the accompanying drawings or may be learned by productionor operation of the examples. The objects and advantages of the presentsubject matter may be realized and attained by means of themethodologies, instrumentalities and combinations particularly pointedout in the appended claims.

BRIEF DESCRIPTION OF THE DRAWINGS

The drawing figures depict one or more implementations, by way ofexample only, not by way of limitations. In the figures, like referencenumerals refer to the same or similar elements.

FIG. 1A is an isometric view showing a cutaway of a pressure vessel thatincludes an in-vessel shield to enhance performance of a nuclear reactorsystem.

FIG. 1B is a cross-sectional view of a nuclear reactor core of thenuclear reactor system.

FIG. 1C illustrates a nuclear reactor system that implements a pressurevessel with an in-vessel shield liner on an interior wall and includinga plurality of control drums.

FIG. 1D illustrates a nuclear reactor system that implements thepressure vessel with the in-vessel shield liner on the interior wall andincluding a plurality of control rods.

FIG. 2 is an overall shielding thickness graph for a reflector andin-vessel shield in eleven different implementations of the nuclearreactor.

FIG. 3 is an example depletion graph of two different types of nuclearreactor cores.

FIG. 4 is an in-vessel shield material optimization chart for thenuclear reactor system.

FIG. 5 is a flowchart depicting an in-vessel shield material selectionmethod.

FIG. 6 is a flowchart depicting an in-vessel shield method that includesin-vessel shield manufacture, as well techniques for in-vessel shielddesign and installation.

Parts Listing 100 Nuclear Reactor System 101 Nuclear Reactor Core 101A-BNuclear Reactor Cores 102A-N Insulator Elements 103A-N ModeratorElements 104A-N Fuel Elements 105 Reflector 105 In-Vessel Shield 107Nuclear Reactor 107A-K Nuclear Reactor Cores 112 Insulator Element Array113 Moderator Element Array 114 Nuclear Fuel Tile Array 115A-N ControlDrums 116 Reflector Material 117 Absorber Material 118A-N Control Rods131A-N In-Vessel Shield Tiles 132 In-Vessel Shield Liner 133 InteriorWall 140 Reflector 141A-N Reflector Blocks 160 Pressure Vessel 200Overall Shielding Thickness Graph 205A-K Percentage of Overall Shieldingas In-Vessel Shield 206A-K Total Thickness 210A-K In-vessel ShieldThickness 211A-K Reflector Thickness 300 Depletion Graph of NuclearReactor Cores 301A-B Nuclear Reactor Core Architectures 305 Lifetime305A-B Lifetimes 310 K-Effective 400 In-Vessel Shield MaterialOptimization Chart 401 Short Lifetime 402 Medium Lifetime 403 LongLifetime 405 Cost 406 Reduction in Pressure Vessel Fast Neutron Fluence411 Short Pareto Front 412 Medium Pareto Front 413 Long Pareto Front 500In-Vessel Shield Material Selection Method 600 In-Vessel Shield Method601 In-Vessel Shield Manufacture 602 In-Vessel Shield Design andInstallation

DETAILED DESCRIPTION

In the following detailed description, numerous specific details are setforth by way of examples in order to provide a thorough understanding ofthe relevant teachings. However, it should be apparent to those skilledin the art that the present teachings may be practiced without suchdetails. In other instances, well known methods, procedures, components,and/or circuitry have been described at a relatively high-level, withoutdetail, in order to avoid unnecessarily obscuring aspects of the presentteachings.

The term “coupled” as used herein refers to any logical or physicalconnection. Unless described otherwise, coupled elements or devices arenot necessarily directly connected to one another and may be separatedby intermediate components, elements, etc.

Unless otherwise stated, any and all measurements, values, ratings,positions, magnitudes, sizes, angles, and other specifications that areset forth in this specification, including in the claims that follow,are approximate, not exact. Such amounts are intended to have areasonable range that is consistent with the functions to which theyrelate and with what is customary in the art to which they pertain. Forexample, unless expressly stated otherwise, a parameter value or thelike may vary by as much as ± 5% or as much as ± 10% from the statedamount. The term “approximately” or “substantially” means that theparameter value or the like varies up to ± 10% from the stated amount.

The orientations of the nuclear reactor core 101, nuclear reactor 107,associated components, and/or any nuclear reactor system 100incorporating the in-vessel shield 105, such as shown in any of thedrawings, are given by way of example only, for illustration anddiscussion purposes. In operation for a particular nuclear reactorsystem 100, the nuclear reactor 107 may be oriented in any otherdirection suitable to the particular application of the nuclear reactor107, for example upright, sideways, or any other orientation. Also, tothe extent used herein, any directional term, such as lateral,longitudinal, up, down, upper, lower, top, bottom, and side, are used byway of example only, and are not limiting as to direction or orientationof any nuclear reactor 107 or component of the nuclear reactor 107constructed as otherwise described herein. Reference now is made indetail to the examples illustrated in the accompanying drawings anddiscussed below.

FIG. 1A is an isometric view showing a cutaway of a pressure vessel 160that includes an in-vessel shield 105 to enhance performance of anuclear reactor system 100. As shown in FIG. 1A a nuclear reactor system100 includes a nuclear reactor 107. Nuclear reactor 107 includes apressure vessel 160 that includes an interior wall 133.

FIG. 1B is a cross-sectional view of a nuclear reactor core 101 of thenuclear reactor system 100. Nuclear reactor system 100 includes thenuclear reactor core 101, and the nuclear reactor core 101 is locatedwithin the interior wall 133 of the pressure vessel 160. Generally, thenuclear reactor core 101 includes a plurality of fuel elements 104A-Nand at least one moderator element 103. In the implementation of FIG.1B, the plurality of fuel elements 104A-N are arranged as a nuclear fueltile array 114 of nuclear fuel tiles 104A-N and the nuclear reactor core101 includes a plurality of moderator elements 103A-N. In a secondexample, the nuclear reactor core 101 can be implemented like thenuclear reactor core 110 described in FIGS. 3-4 and the associated textof U.S. Pat. No. 10,643,754 to Ultra Safe Nuclear Corporation ofSeattle, Washington, issued May 5, 2020, titled “Passive ReactivityControl of Nuclear Thermal Propulsion Reactors” the entirety of which isincorporated by reference herein. In the second example, the fuelelements 104A-N can be implemented like the fuel elements 310A-N and themoderator elements 103A-N can be implemented like the tie tubes 320A-Ndescribed in FIGS. 3-4 and the associated text of U.S. Pat. No.10,643,754.

In a third example, the nuclear reactor core 101 can be implemented likethe nuclear reactor core 101 described in FIG. 2C and the associatedtext of U.S. Pat. Pub. No. 2020/0027587 to Ultra Safe NuclearCorporation of Seattle, Washington, published Jan. 23, 2020, titled“Composite Moderator for Nuclear Reactor Systems,” the entirety of whichis incorporated by reference herein. In the third example, the fuelelements 104A-N can be implemented like the fuel elements 102A-N and themoderator elements 103A-N can be implemented like the compositemoderator blocks described in FIG. 2C and the associated text of U.S.Pat. Pub. No. 2020/0027587.

As further shown in FIG. 1B, nuclear reactor 107 includes a reflector140 (e.g., an outer reflector region) located inside the pressure vessel160. Reflector 140 includes a plurality of reflector blocks 141A-Nlaterally surrounding the plurality of fuel elements 104A-N and the atleast one moderator element 103. As shown in FIG. 1A, nuclear reactor107 includes an in-vessel shield 105 located on the interior wall 133 ofthe pressure vessel 160 to surround the reflector blocks 141A-N.

In-vessel shield 105 is formed of two or more neutron absorbingmaterials. The two more neutron absorbing materials include a near blackneutron absorbing material and a gray neutron absorbing material. Nearblack neutron absorbing material includes a composited ceramic material.Gray neutron absorbing material includes a heavy metal material. Thecomposited ceramic material and the heavy metal material form thein-vessel shield. Composited ceramic material includes boron carbide(B₄C), hafnium carbide (HfC), or gadolinium oxide (Gd₂O₃). Compositedceramic material can further include aluminum oxide (Al₂O₃) or siliconcarbide (SiC). Heavy metal material includes tungsten (W), iron (Fe),Nickel (Ni), or copper (Cu).

More specifically, composited ceramic material can include: boron-10carbide (10B₄C), a boron-10 carbide and aluminum oxide composite(¹⁰B₄C—Al₂O₃) of 50% boron-10 carbide by weight percent, a boron-10carbide and silicon carbide composite (¹⁰B₄C—SiC) of 50% boron-10carbide by weight percent, or a borated stainless steel alloy of 5%boron-10 by weight percent. The heavy metal material includes atomizedtungsten heavy metal with a tungsten content greater than or equal to90% by weight percent. Composited ceramic material can include aplurality of composited ceramic particles with an average particlediameter greater than or equal to approximately 80 nanometers and lessthan or equal to approximately 100 microns. The plurality of compositedceramic particles can be embedded inside a heavy metal matrix of theheavy metal material to form the in-vessel shield 105.

In-vessel shield 105 can be formed integrally (e.g., as one component,part, or piece) as an in-vessel shield liner 132 that is a suitableshape to line the interior wall 133, e.g., a tube or pipe shape.Alternatively, the in-vessel shield 105 is formed as several componentsformed separately and then connected together. Hence, as depicted inFIG. 1B, in-vessel shield 107 is formed as a plurality of in-vesselshield tiles 131A-N, that are joined together to form the in-vesselshield liner 132. The plurality of in-vessel shield tiles 131A-N aredisposed on (e.g., attached to) the interior wall 133 to form thein-vessel shield 105.

In-vessel shield tiles 131A-N are tolerant to swelling and reduceneutron streaming paths. In-vessel shield tiles 131A-N have acomposition and geometry that can be designed particularly for thenuclear reactor system 100, for example, the plurality of in-vesselshield tiles 131A-N include a base shape with an interlocking geometrypattern. In FIG. 1A, the in-vessel shield 105 is made up of thein-vessel shield tiles 131A-N with the base shape. All or a subset ofthe plurality of in-vessel shield tiles 131A-N can be a curvedpolyhedron shape or a truncated portion thereof to match the curvatureof the interior wall 133. In-vessel shield tiles 131A-N are also shapedto conform to match contour(s) (e.g., curvature) of the surface(s) ofthe interior wall 133. In FIG. 1B, the in-vessel shield tiles 131A-N areshaped as a curved hexagonal prism in three-dimensional space or acurved hexagon in two-dimensional space. In-vessel shield tiles 131A-Ncan be another polyhedron shape (e.g., a triangular prism or a cuboid)in three-dimensional space or circular, oval, square, rectangular,triangular, or another polygon shape in two-dimensional space.

Interior wall 133 can be formed of a continuous surface, e.g., rounded,aspherical, or spherical surfaces to form a cylinder or other conicalsurfaces to form a quadric surface, such as a hyperboloid, cone,ellipsoid, paraboloid, etc. Alternatively or additionally, the interiorwall 133 can be formed of a plurality of discontinuous surfaces (e.g.,to form a cuboid or other polyhedron). As used herein, “discontinuous”means that the surfaces in aggregate do not form a continuous round(e.g., circular or oval) perimeter of the interior wall 133. In FIG. 1A,the portion of the interior wall 133 shown is a rounded continuoussurface. The interlocking geometry pattern of the plurality of in-vesselshield tiles 131A-N are joined to cover and collectively form anin-vessel shield liner 132 on the continuous or discontinuous surfacesof the interior wall 133.

Returning now to FIG. 1B, nuclear reactor 107 includes the nuclearreactor core 101, in which a controlled nuclear chain reactions occurs,and energy is released. The neutron chain reaction in the nuclearreactor core 101 is critical — a single neutron from each fissionnucleus results in fission of another nucleus — the chain reaction mustbe controlled. By sustaining controlled nuclear fission, the nuclearreactor system 100 produces heat energy. In an example implementation,the nuclear reactor system 100 is implemented as a gas-cooled hightemperature nuclear reactor 107. However, the nuclear reactor system 100with the in-vessel shield 105 can enable breakthrough performance byreducing mass and size and improving power per mass in a large utilityscale nuclear reactor, heat pipe nuclear reactor, molten-salt-coolednuclear reactor, fuel-in-salt nuclear reactor, or a sodium-cooled fastnuclear reactor. For example, in-vessel shield 105 can be included in anuclear reactor system 100, such as a gas-cooled graphite-moderatednuclear reactor, a fluoride salt-cooled high-temperature nuclear reactorwith a higher thermal neutron flux than the gas-cooledgraphite-moderated nuclear reactor, or a sodium fast nuclear reactorwith a faster neutron flux than the gas-cooled graphite-moderatednuclear reactor.

In the depicted example, the nuclear reactor system 100 with the nuclearreactor core 101 is utilized in a space environment, such as in anuclear thermal propulsion (NTP) system. An example NTP system that theintegrated in-vessel shield 105 can be implemented in is described inFIGS. 1-2 and the associated text of U.S. Pat. No. 10,643,754 to UltraSafe Nuclear Corporation of Seattle, Washington, issued May 5, 2020,titled “Passive Reactivity Control of Nuclear Thermal PropulsionReactors” the entirety of which is incorporated by reference herein. Forexample, the nuclear reactor system 100 that includes the in-vesselshield 105 can be a nuclear thermal rocket reactor, nuclear electricpropulsion reactor, Martian surface reactor, or lunar surface reactor.

In such an NTP system (e.g., compact space nuclear reactor), a generatedthrust propels a vehicle that houses, is formed integrally with,connects, or attaches to the nuclear reactor core 101, such as a rocket,drone, unmanned air vehicle (UAV), aircraft, spacecraft, missile, etc.Typically, this is done by heating a propellant, typically low molecularweight hydrogen, to over 2,600° Kelvin by harnessing thermal energy fromthe nuclear reactor core 101. In addition, the NTP nuclear reactorsystem 100 can be used in the propulsion of submarines or ships.

As noted above, the nuclear reactor system 100 can also be a nuclearpower plant in a terrestrial land application, e.g., for providingnuclear power (e.g., thermal and/or electrical power) for remote regionapplications, including outer space, celestial bodies, planetary bodies,and remotes regions on Earth. An example terrestrial land nuclearreactor system that the in-vessel shield 105 can be implemented in isdescribed in FIG. 1 and the associated text of U.S. Pat. Pub. No.2020/0027587 to Ultra Safe Nuclear Corporation of Seattle, Washington,published Jan. 23, 2020, titled “Composite Moderator for Nuclear ReactorSystems,” the entirety of which is incorporated by reference herein. Forexample, the nuclear reactor system 100 can be a small commercialfission power system for near term space operations, lunar landers, or acommercial fission power system for high-power spacecraft andlarge-scale surface operations, such as in-situ resource utilization. Inanother example, the nuclear reactor system 100 with the in-vesselshield 105 is utilized in a space reactor for electrical powerproduction on a planetary surface.

Nuclear reactor system 100 can also be a terrestrial power system, suchas a nuclear electric propulsion (NEP) system for fission surface power(FSP) system. NEP powers electric thrusters such as a Hall-effectthruster for robotic and human spacecraft. FSP provides power forplanetary bodies such as the moon and Mars. In the NEP and FSP powerapplications, the nuclear reactor system 100 heats a working fluid(e.g., He, HeXe, Ne, CO₂) through a power conversion system (e.g.,Brayton) to produce electricity. Moreover, in the NEP and FSP powerapplications, the nuclear reactor system 100 does not include apropellant, but rather includes a working fluid that passes through areactor inlet when producing power. In the NEP and FSP powerapplications, the moderator elements 103A-N can be cooled via thereactor inlet working fluid (e.g., the flow coming out of a recuperator)before the working fluid passes through the fuel elements 104A-N.

Each of the fuel elements 104A-N, shown as nuclear fuel tiles 104A-N,includes a nuclear fuel. The nuclear fuel includes a fuel compactcomprised of coated fuel particles, such as tristructural-isotropic(TRISO) fuel particles embedded inside a high-temperature matrix. Insome implementations, the nuclear fuel includes a fuel compact comprisedof bistructuralisotropic (BISO) fuel particles embedded inside thehigh-temperature matrix. The high-temperature matrix includes siliconcarbide, zirconium carbide, titanium carbide, niobium carbide, tungsten,molybdenum, or a combination thereof. Each of the TRISO fuel particlescan include a fuel kernel surrounded by a porous carbon buffer layer, aninner pyrolytic carbon layer, a binary carbide layer (e.g., ceramiclayer of SiC or a refractory metal carbide layer), and an outerpyrolytic carbon layer. The refractory metal carbide layer of the TRISOfuel particles can include at least one of titanium carbide (TiC),zirconium carbide (ZrC), niobium carbide (NbC), tantalum carbide,hafnium carbide, ZrC—ZrB₂ composite, ZrC—ZrB₂—SiC composite, or acombination thereof. The high-temperature matrix can be formed of thesame material as the binary carbide layer of the TRISO fuel particles.

A description of TRISO fuel particles dispersed in a silicon carbidematrix to form a cylindrical shaped nuclear fuel compact is provided inthe following patents and publications of Ultra Safe Nuclear Corporationof Seattle, Washington: U.S. Pat. No. 9,299,464, issued Mar. 29, 2016,titled “Fully Ceramic Nuclear fuel and Related Methods”; U.S. Pat. No.10,032,528, issued Jul. 24, 2018, titled “Fully CeramicMicro-encapsulated (FCM) fuel for CANDUs and Other Reactors”; U.S. Pat.No. 10,109,378, issued Oct. 23, 2018, titled “Method for Fabrication ofFully Ceramic Microencapsulation Nuclear Fuel”; U.S. Pat. Nos. US9,620,248, issued Apr. 11, 2017 and 10,475,543, issued Nov. 12, 2019,titled “Dispersion Ceramic Micro-encapsulated (DCM) Nuclear Fuel andRelated Methods”; U.S. Pat. Pub. No. 2020/0027587, published Jan. 23,2020, titled “Composite Moderator for Nuclear Reactor Systems”; and U.S.Pat. No. 10,573,416, issued Feb. 25, 2020, titled “Nuclear Fuel ParticleHaving a Pressure Vessel Comprising Layers of Pyrolytic Graphite andSilicon Carbide,” the entireties of which are incorporated by referenceherein. As described in those Ultra Safe Nuclear Corporation patents,the nuclear fuel can include a cylindrical fuel compact or pelletcomprised of TRISO fuel particles embedded inside a silicon carbidematrix to create a cylindrical shaped nuclear fuel compact.

As shown, nuclear reactor core 101 includes an insulator element array112 of insulator elements 102A-N and a moderator element array 113 ofmoderator elements 103A-N. Insulator elements 102A-N are formed of ahigh-temperature thermal insulator material with low thermalconductivity. The high-temperature thermal insulator material caninclude low density carbides, metal-carbides, metal-oxides, or acombination thereof. More specifically, the high-temperature thermalinsulator material includes low density SiC, stabilized zirconium oxide,aluminum oxide, low density ZrC, low density carbon, or a combinationthereof. Moderator elements 103A-N are formed of a low-temperaturesolid-phase moderator. The low-temperature solid-phase moderatorincludes MgH_(x), YH_(x), ZrH_(x), CaH_(x), ZrO_(x), CaO_(x), BeO_(x),BeC_(x), Be, enriched boron carbide, ¹¹B₄C, CeH_(x), LiH_(x), or acombination thereof.

In an NTP, NEP, or FSP nuclear reactor system 100, the nuclear reactor107 can include a plurality of control drums 115A-N and a reflector 140.The control drums 115A-N may laterally surround the insulator elementarray 112 of insulator elements 102A-N, the moderator element array 113of moderator elements 103A-N, and nuclear fuel tile array 114 of nuclearfuel tiles 104A-N to change reactivity of the nuclear reactor core 101by rotating the control drums 115A-N. As depicted, the control drums115A-N reside on the perimeter or periphery of a pressure vessel 160 andare positioned circumferentially around the insulator elements 102A-N,moderator elements 103A-N, and nuclear fuel tiles 104A-N of the nuclearreactor core 101. Control drums 115A-N may be located in an area of thereflector 140, e.g., an outer reflector region formed of reflectorblocks 141A-N immediately surrounding the nuclear reactor core 101, toselectively regulate the neutron population and nuclear reactor powerlevel during operation. For example, the control drums 115A-N can be acylindrical shape and formed of both a reflector material 116 (e.g.,beryllium (Be), beryllium oxide (BeO), BeSiC, BeMgO, Al₂O₃, etc.) on afirst outer surface and an absorber material 117 on a second outersurface.

The reflector material 116 and the absorber material 117 can be onopposing sides of the cylindrical shape, e.g., portions of an outercircumference, of the control drums 115A-N. The reflector material 116can include a reflector substrate shaped as a cylinder or a truncatedportion thereof. The absorber material 117 can include an absorber plateor an absorber coating. The absorber plate or the absorber coating aredisposed on the reflector substrate to form the cylindrical shape ofeach of the control drums 115A-N. For example, the absorber plate or theabsorber coating covers the reflector substrate formed of the reflectormaterial to form the control drums 115A-N.

Rotating the depicted cylindrical-shaped control drums 115A-N changesproximity of the absorber material 117 (e.g., boron carbide, B₄C) of thecontrol drums 115A-N to the nuclear reactor core 101 to alter the amountof neutron reflection. When the reflector material 116 is inwards facingtowards the nuclear reactor core 101 and the absorber material 117 isoutwards facing, neutrons are scattered back (reflected) into thenuclear reactor core 101 to cause more fissions and increase reactivityof the nuclear reactor core 101. When the absorber material 117 isinwards facing towards the nuclear reactor core 101 and the reflectormaterial 116 is outwards facing, neutrons are absorbed and furtherfissions are stopped to decrease reactivity of the nuclear reactor core101. In a terrestrial land application, the nuclear reactor core 101 mayinclude control rods 118A-N (see FIG. 1D) composed of chemical elementssuch as boron, silver, indium, and cadmium that are capable of absorbingmany neutrons without themselves fissioning.

Neutron reflector 140, e.g., shown as the outer reflector region, can befiller elements disposed between outermost nuclear fuel tiles 104A-N andthe control drums 115A-N as well as around the control drums 115A-N.Reflector 140 can be formed of a moderator that is disposed between theoutermost nuclear fuel tiles 104A-N and an optional barrel (e.g., formedof beryllium). The reflector 140 can include hexagonal or partiallyhexagonal shaped filler elements and can be formed of a neutronmoderator (e.g., beryllium oxide, BeO). Although not required, nuclearreactor 107 can include the optional barrel (not shown) to surround thebundled collection that includes the insulator element array 112,moderator element array 113, nuclear fuel tile array 114 of the nuclearreactor core 101, as well as the reflector 140. As depicted, the controldrums 115A-N reside on the perimeter of the pressure vessel 160 and canbe interspersed or disposed within the reflector 140, e.g., surround asubset of the filler elements (e.g., reflector blocks 141A-N) formingthe reflector 140.

Pressure vessel 160 can be formed of aluminum alloy, carbon-composite,titanium alloy, a radiation resilient SiC composite, nickel based alloys(e.g., Inconel™ or Haynes™), or a combination thereof. Pressure vessel160 and nuclear reactor system 100 can be comprised of other components,including cylinders, piping, and storage tanks that transfer a moderatorcoolant that flows through moderator coolant passages 121A-N; and aseparate nuclear fuel coolant, such as a propellant (e.g., hydrogen gasor liquid) that flows through the fuel coolant passages 141A-N. Themoderator coolant and the nuclear fuel coolant can be a gas or a liquid,e.g., that transitions from a liquid to a gas state during a burn cycleof the nuclear reactor core 101 for thrust generation in an NTP nuclearreactor system 100. Hydrogen is for an NTP nuclear reactor system 100.In NEP or FSP applications, the nuclear reactor system 100 circulates aworking fluid, such as He, neon, HeXe, CO₂, instead.

In the example of FIG. 1B, nuclear reactor system 100 advantageouslyenables the moderator coolant to flow through the moderator coolantpassages 121A-N and a separate nuclear fuel coolant (e.g., a propellant,such as hydrogen gas) to flow through the fuel coolant passages 141A-N.The moderator coolant passages 121A-N are flattened ring shaped (e.g.,O-shape) openings, such as a channels or holes to allow the moderatorcoolant to pass through in the nuclear reactor core 101 and into a heatsink (not shown) via a dedicated moderator coolant loop, for example.The fuel coolant passages 141A-N are channels or holes to allow thenuclear fuel coolant to pass through in the nuclear reactor core 101 andinto a thrust chamber (not shown) for propulsion in a separate nuclearfuel coolant loop, for example.

In an alternative implementation, a coolant that is shared between themoderator elements 103A-N and the nuclear fuel tiles 104A-N may beflowed through both the moderator coolant passages 121A-N and the fuelcoolant passages 141A-N. In the alternative implementation, the coolantthat flows through the plurality of fuel elements 104A-N can includehelium, FLiBe molten salt formed of lithium fluoride (LiF), berylliumfluoride (BeF₂), sodium, He, HeXe, CO₂, neon, or HeN. The shared coolantflows through the moderator coolant passages 121A-N before the sharedcoolant is heated in the nuclear fuel tiles 104A-N. This keeps themoderator elements 103A-N cool.

FIG. 1C illustrates a nuclear reactor system 100 that implements apressure vessel 160 with the in-vessel shield liner 132 on the interiorwall 133 and including control drums 115A-N. FIG. 1D illustrates anuclear reactor system 100 that implements the pressure vessel 160 withthe in-vessel shield liner 132 on the interior wall 133 and includingcontrol rods 118AN. Control rods 118A-N may be positioned in an area ofthe nuclear reactor core 101 to regulate the neutron population andnuclear reactor power level during operation by changing reactivity ofthe nuclear reactor core 101. Control rods 118A-N protrude from the topof the pressure vessel 160, can insert the length of the nuclear reactor107, but also have the ability to be withdrawn from the nuclear reactor107. Control drums 115A-N enable horizontal, axially symmetricalconfigurations for the nuclear reactor 107, which can be more easilytransported and deployed. Control drums 115A-N also maximize the usablevolume in the nuclear reactor core 101, as they can be integrated intothe reflector 140, while control rods 118A-N leave an un-fueled void inthe nuclear reactor core 101 when withdrawn.

Reflector 140 may be an integrally formed body (e.g. a tube or pipe), tosurround the nuclear reactor core 101, or it may be several componentsor parts, such as a reflector region made of reflector blocks 141A-N,which surround the nuclear reactor core 101. Reflector 140 elasticallyscatters neutrons attempting to escape the nuclear reactor core 101, andredirects these neutrons back toward the nuclear reactor core 101 inorder to create more fission events and therefore more energy. Reflector140 also has a secondary purpose: by redirecting neutrons back towardthe nuclear reactor core 101, the reflector 140 necessarily reduces theamount of neutrons impacting the pressure vessel 160 and areas beyond.Reflector 140 reduces the fast neutron fluence or fast neutron flux, thetotal length travelled by all free fast neutrons per time and volumeoutside the nuclear reactor core 101. Efficacy of the reflector 140 inboth the electron scattering and the fast neutron fluence reduction islogarithmically improved by increasing the thickness of the reflector140.

In a conventional nuclear reactor, the thickness of the reflector 140 isincreased until the fast neutron fluence outside the reflector 140 isacceptably low: this thickness is often materially beyond the thicknessrequired to efficiently reflect neutrons back toward the nuclear reactorcore 101. In contrast, the nuclear reactor system 100 described hereinadditionally includes in-vessel shield 105, which can behave as aneutron poison, and is designed to absorb free neutrons, rather thanreflect free neutrons back toward the nuclear reactor core 101.In-vessel shield 105 can absorb free neutrons to stop fast neutronfluence through the neutron poison. In-vessel shield 105, while beingsignificantly thinner than the reflector 140, can nevertheless reducethe fast neutron fluence in an area by the same amount as the reflector140. Although the in-vessel shield 105 can have the effect of neutronpoisoning, the in-vessel shield 105 does not behave as a pure neutronpoison because the neutron poison is composited within a moderator toform the in-vessel shield 105. By using a combination of materials withlarge neutron absorption cross-sections as in-vessel shield 105 betweenthe pressure vessel 160 and the reflector 140 of the nuclear reactor107, the fast neutron fluence can be reduced to acceptable levels.

Pressure vessel 160 includes the reflector 140 disposed therein, for theprimary purpose of redirecting neutrons back to the nuclear reactor core101, and secondarily for the purpose of reducing fast neutron fluence toan acceptable level at and beyond the pressure vessel 160. Pressurevessel 160 is then lined with in-vessel shield 105, with the primarypurpose of reducing fast neutron fluence to an acceptable level at andbeyond the pressure vessel 160. The in-vessel shield 105 is disposedbetween the pressure vessel 160 and the reflector 140. Together, thein-vessel shield 105 and the reflector 140 are able to reduce fastneutron fluence to an acceptable level at and beyond the pressure vessel160, while the combined thickness of the in-vessel shield 105 and thereflector 140 is overall thinner than an equivalent reflector 140without a paired in-vessel shield 105 reducing fast neutron fluence tothe same acceptable or equivalent level. Overall, this allows for acombined thinner reflector 140 and in-vessel shield 105 than a reflector140 alone, and a smaller pressure vessel 160 and nuclear reactor 107with a reduced volume and mass.

Because of fast neutron fluence concerns, any size nuclear reactorsystem 100 (large or small) can benefit from the in-vessel shield 105.In a large nuclear reactor system 100, the fast neutron fluence to thepressure vessel 160 is limited by the distance of the active nuclearreactor core 101 from the pressure vessel 160. For a volume-constrainedsmall nuclear reactor system 100, the use of in-vessel shield 105 can bevery advantageous. In-vessel shield 105 reduces the radiation fieldaround the nuclear reactor system 100, which in turn decreases theactivation in the area outside (e.g., around) the nuclear reactor 107,decreases the footprint, and aids in ease of install.

During operation of the nuclear reactor system 100, the materialsdisposed within the pressure vessel 160 become activation products, maderadioactive by the free neutrons. Nuclear reactor system 100 includes aninner density of activation product within the nuclear reactor core 101and an outer density of activation product outside the nuclear reactorcore 101. The outer density of activation product is lower than theinner density of activation product during operation of the nuclearreactor core 101. This is because the purpose of the reflector 140 is toredirect free neutrons back into the nuclear reactor core 101,increasing core activation. The in-vessel shield 105 is made ofmaterials that are selected because the materials absorb free neutronswhile becoming minimally radioactive: this further reduces activationoutside the nuclear reactor core 101, in comparison to the activationwithin the nuclear reactor core 101.

In-vessel shield 105 is formed out of two or more neutron absorbingmaterials, at least one of which is a near black neutron absorbingmaterial, and another is a gray neutron absorbing material. Near blackneutron absorbing material is more efficient at absorbing free neutronsand reducing neutron fluence than a gray neutron absorbing material insimilar quantities. Near black neutron absorbing material is acomposited ceramic material, that is, a composited high-temperatureneutron absorbing ceramic. For example, the composited ceramic materialincludes mixtures of neutron absorbing ceramics, such as B₄C, HfC, andGd₂O₃ with radiation tolerant materials, such as Al₂O₃ and SiC, tominimize radiation induced swelling. Adding radiation tolerantmaterials, such as Al₂O₃ and SiC, to form the composted ceramic materialincreases temperature and radiation endurance of the near black neutronabsorbing material.

The gray neutron absorbing material includes a heavy metal. Together,the near black neutron material and the gray neutron absorbing materialform an in-vessel shield 105. In-vessel shield 105 is fabricated throughadvanced manufacturing of the composited materials specificallyengineered for shielding. This contrasts with the current paradigm ofshield selection based on selection of the generic class of engineeringalloys and structural ceramics, perhaps with some degree of isotopicenrichment. In-vessel shield 105 may be primarily made of compositedceramics, and therefore could also be described in such examples as acomposited ceramic shield.

FIG. 2 is an overall shielding thickness graph 200 for a reflector 140and in-vessel shield 105 in eleven different implementations of thenuclear reactor 107A-K. Overall shielding graph 200 shows that areduction of the total thickness 206A-K of the nuclear reactor 107 isachieved by: (a) increasing the in-vessel shield thickness 210A-K; and(b) reducing the reflector thickness 211A-K. In other words, increasingthe percentage of overall shielding as in-vessel shield 205A-K directlyreduces the total thickness 206A-K of the nuclear reactor 107.Accordingly, the in-vessel shield 105 technology can improve performanceof the nuclear reactor system 100 by enabling the nuclear to be lighterand more compact.

In the overall shielding thickness graph 200, white bars represent thein-vessel shield thickness 210A-K added by the in-vessel shield 105,while the black bars represent the reflector thickness 211A-K added bythe reflector 140. Total thickness 206A-K is the sum of the in-vesselshield thickness 210A-K and the reflector thickness 211A-K. As shown, ina first implementation of nuclear reactor 107A, nuclear reactor 107Aincludes a total thickness 206A of 30 centimeters (cm), where thereflector thickness 211A is 30 cm and the in-vessel shield thickness210K is 0 cm. In the eleventh implementation of nuclear reactor 107K,nuclear reactor 107K includes a total thickness 206K of 10 cm, where thereflector thickness 211K is 0 cm and the in-vessel shield thickness 210Kis 10 cm.

Based on the total thicknesses 206A-K of the overall shielding thicknessgraph 200, in-vessel shield 105 has a shield diffusion length threetimes shorter than a reflector diffusion length of the reflector 140.Diffusion length is in part a factor of the diffusion coefficient of thematerial, as well as the macroscopic absorption cross section of thatmaterial. This diffusion length difference of three between thereflector 140 and in-vessel shield 105 is based on differences in thematerials forming the in-vessel shield 105 and the reflector 140.Diffusion length difference may vary, e.g., based on materials selected,differences in the in-vessel shield thickness 210A-K, reflectorthickness 211A-K, and overall geometry of the nuclear reactor core 101.

In FIG. 2 , the total thickness 206A-K represents the same overallamount of neutron fluence reduction. As shown, a 10 cm in-vessel shieldthickness 210K for the in-vessel shield 105 results in the same neutronfluence reduction as a 30 cm reflector thickness 211A for the reflector140. Advantageously, the 10 cm in-vessel shield thickness 210K for thein-vessel shield 105 achieves a pressure vessel 160 with a 20 cm smallerradius, which is a 66% reduction in total thickness 206K. Moreover, thenuclear reactor 107K can be lighter, as the 10 cm of in-vessel shieldthickness 210K may weigh less than the equivalent 30 cm of reflectorthickness 211A, and in-vessel shield 105 equivalently reduces fastneutron fluence at and beyond the pressure vessel 160.

A second implementation of nuclear reactor 107B has 1 cm of in-vesselshield thickness 210B, which reduces total thickness 206B by 2 cmcompared to the first implementation of the nuclear reactor 107A whereno in-vessel shield 105 is present. Every 1 cm of in-vessel shieldthickness 210 reduces the need for 3 cm of reflector thickness 211.Based on the overall shielding thickness graph 200, an optimal balancecan be met: if only 15 cm reflector thickness 211F is needed toefficiently reflect free neutrons back toward the nuclear reactor core101, then the sixth implementation of nuclear reactor 107F shows thatonly 5 cm in-vessel shield thickness 210F is needed to efficientlyreduce fast neutron fluence to an acceptable level at and beyond thepressure vessel 160.

In a conventional nuclear reactor without the in-vessel shield 105, thepressure vessel 160 needs 30 cm of additional radius to properly reflectfree neutrons back toward the nuclear reactor core 101 and toefficiently reduce fast neutron fluence to an acceptable level at andbeyond the pressure vessel 160. Advantageously, the nuclear reactorsystem 100 with the in-vessel shield 105 only requires 20 cm ofadditional radius to properly reflect free neutrons back toward thenuclear reactor core 101 and to efficiently reduce fast neutron fluenceto an acceptable level at and beyond the pressure vessel 160. An evenfurther improvement in reduction in size and mass can be achieved if thenuclear reactor system 100 includes control rods 118A-N in a drum formfactor to alternatively face reflector blocks 141A-N of the reflector140 and moderator elements 103A-N of nuclear reactor core 101. In such anuclear reactor system 100 with control rods 118A-N, the pressure vessel160 does not necessarily require a fully efficient reflector 140 toreflect free neutrons back toward the nuclear reactor core 101.

Accordingly, overall shielding thickness graph 200 of FIG. 2 depicts howto design a nuclear reactor system 100, where a reflector block 141 ofthe plurality of reflector blocks 141A-N has a reflector thickness 211,a reflector diffusion coefficient based on a reflector material thatforms the reflector block 141, and a reflector macroscopic absorptioncross section based on the reflector material. Additionally, thein-vessel shield 105 has an in-vessel shield thickness 210, a shielddiffusion coefficient based on an in-vessel shield material that formsthe in-vessel shield 105, and an in-vessel shield macroscopic absorptioncross section based on the in-vessel shield material. The reflectorblock 140 in combination with the in-vessel shield 105 have a combineddiffusion length, based at least on: the reflector thickness 211, thereflector diffusion coefficient based on the reflector material, thereflector macroscopic absorption cross section based on the reflectormaterial, the in-vessel shield thickness 210, the shield diffusioncoefficient based on the in-vessel shield material, and the shieldmacroscopic absorption cross section based on the in-vessel shieldmaterial. In an example efficient nuclear reactor system 100, thereflector thickness 211 and the in-vessel shield thickness 210 whenadded together are less (e.g., not thicker) than double the combineddiffusion length. This is because the performance of a reflector 140does not materially improve when the reflector 140 is thicker thandouble the diffusion length of the reflector material. Therefore, thenuclear reactor system 100 including both the reflector 140 andin-vessel shield 105 can provide the same amount of diffusion as aconventional nuclear reactor system that includes the reflector 140 (andno in-vessel shield 105), but provide the equivalent diffusion in a muchsmaller amount of space.

FIG. 3 is an example depletion graph 300 of two different types ofnuclear reactor cores 101A-B. As shown, the coefficient k-effective(k-eff) 310 over the lifetime 305 as measured in years of the nuclearreactor cores 101A-B is improved. K-eff 310, also known as the neutronmultiplication factor, characterizes the criticality state of thefissile material in the nuclear reactor core 101. Generally K-eff =number of neutrons produced / number of neutrons lost (through leakageor absorption). If K-eff 310 is greater than or equal to 1, only thencan the nuclear fission chain reaction can be sustained. As shown, afirst reactor nuclear core architecture 301A enables a nuclear reactorcore 101A with a lifetime 305A of approximately 10 years. A secondnuclear reactor core architecture 301B enables a nuclear reactor core101B with a lifetime 305B of approximately 15 years. Based on theexpected lifetimes 305A-B of the nuclear reactor cores 101A-B, anappropriate amount and type of in-vessel shield 105 and reflector 140are selected. Even if the in-vessel shield 105 or reflector 140 is anon-burnable neutron poison, ultimately all neutron poisons degrade inefficacy as they experience neutron fluence for an extended period oftime. Thus, the first nuclear reactor core 301A potentially may requirea thicker in-vessel shield 105 than the second nuclear reactor core301B, e.g., assuming that any effect of geometrical buckling is similarin both nuclear reactor core architectures 301A-B.

FIG. 4 is an in-vessel shield material optimization chart 400 for thenuclear reactor system 100. In-vessel shield material optimization chart400 shows three Pareto fronts 411-413, including a short Pareto front411, a medium Pareto front 412, and a long Pareto front 413. Each Paretofront 411-413 is divided between different in-vessel shield materialswith a short lifetime 401, in-vessel shield materials with a mediumlifetime 402, and in-vessel shield materials with a long lifetime 403.

As explained in FIG. 3 , lifetime 305 (which correlates to K-effective305) is how long the nuclear reactor system 100 can operate. In FIG. 4 ,the three lifetimes 401-403 are assumed to be the most common initialselection criteria of the in-vessel shield material optimization chart400. A short lifetime 401, a medium lifetime 402, and a long lifetime403 determine the number of Pareto fronts 411-413. Pareto fronts 411-413are utilized to evaluate design options of the nuclear reactor system100 and the design options are on a Pareto front 411-413 if the designoption scores the highest on all figures of merit possible withoutdecreasing other figures of merit. Therefore, an in-vessel shieldmaterial on the short Pareto front 411 at a given cost 405 will have thegreatest reduction in pressure vessel fast neutron fluence 406 for thenuclear reactor system 100 with a short lifetime 401.

Three Pareto fronts 411-413 of FIG. 4 demonstrate how different neutronpoisons can fit in the design of the in-vessel shield 105. First, asimplistic B₄C material effectively reduces fast neutron fluence at aminimal cost. Boron-10 within B₄C (also known as ¹⁰B₄C) has a largeabsorption cross section of fast spectrum neutrons. This means that anin-vessel shield 105 formed of B₄C does not need to be as thick asanother in-vessel shield 105 formed of another material with a smallerabsorption cross section of fast spectrum neutrons. Pure B₄C has a shortlifetime because of helium-induced swelling that B₄C undergoes in aradiation environment. Therefore, pure B₄C material is on the line ofthe short Pareto front 411.

Second, B₄C—Al₂O₃ composite or B₄C—SiC composites can be on the order of50% B₄C by weight and better accommodate the swelling of B₄C. B₄C—Al₂O₃composite or B₄C—SiC composites are less effective in shielding fastneutrons than pure B₄C. Thus, B₄C—Al₂O₃ composite or B₄C—SiC compositesare on the line of the medium Pareto front 412.

Third, borated stainless steel is less than 5% boron by weight, but isvery dimensionally-stable under irradiation. Borated stainless steel ison the line of the long Pareto front 413. Based on the in-vessel shieldmaterial optimization chart 400, the two or more neutron absorbingmaterials to form the in-vessel shield 105 are selected, including anear black neutron absorbing material and a gray neutron absorbingmaterial.

FIG. 5 is a flowchart depicting an in-vessel shield material selectionmethod 500. This example assumes that the dependent variable is materialcost, and the independent variables are lifetime 305 of the nuclearreactor system 100, reflector thickness 211, an amount of fast neutronfluence reduced by the reflector 140, and a maximum radius of thepressure vessel 160. Any of these independent variables could beswitched for the dependent variable: for example, if a nuclear reactorsystem 100 already has a pressure vessel 160 and has been retrofittedwith in-vessel shield 105, then the material selection method 500 candetermine the reflector thickness 211 of the reflector 140 needed forthe nuclear reactor system 100 to run efficiently. In-vessel shieldmaterial selection method 500 assumes that the radius of the nuclearreactor core 101 and the fast neutron fluence produced by the nuclearreactor core 101 are known values.

In block 510, the in-vessel shield material selection method 500includes determining the lifetime 305 of the nuclear reactor system 100.Moving to block 520, the in-vessel shield material selection method 500further includes determining a reflector thickness 211 that is efficientbased on: (i) the nuclear reactor core 101; (ii) a reflector materialutilized to form the reflector 140; (iii) geometry of the pressurevessel 160; or (iv) a combination thereof. Ultimately, a reflectorthickness 211 is determined that most efficiently redirects neutronsback to the nuclear reactor core 101.

Continuing to block 530, the in-vessel shield material selection method500 includes determining the reduction of in vessel fast neutron fluenceat and beyond the pressure vessel 160 caused by the reflector 140. Anyamount of fast neutron fluence reduction by the reflector 140 reducesthe amount of fast neutron fluence reduction required by the in-vesselshield 105 to preserve the targeted reduction in pressure vessel 160fast neutron fluence for the nuclear reactor system 100. In block 540,the in-vessel shield material selection method 500 includes determininga maximum pressure vessel radius of the pressure vessel 160 based on:(i) installation; (ii) transportation; and/or (iii) footprintrequirements of the nuclear reactor system 100, e.g., a nuclear reactorcore radius of the nuclear reactor core 101.

In block 550, the in-vessel shield material selection method 500includes selecting the in-vessel shield material of the in-vesselshield. Referring to the in-vessel shield material optimization chart400 of FIG. 4 , the proper Pareto front 411-413 can be selected based onthe short lifetime 401, medium lifetime 402, and long lifetime 403 (seeblock 510). The reduction in pressure vessel fast neutron fluence isdetermined by two factors: (i) a reduction in fast neutron fluencerequirement of the pressure vessel 160; and (b) a maximum thickness ofthe in-vessel shield 105. The reduction in fast neutron fluencerequirement of the pressure vessel 160 is the difference between thefast neutron fluence created by the nuclear reactor core 101 and thefast neutron fluence reduction due to the reflector 140 from block 530.The maximum thickness of the in-vessel shield 105 is the radius of thepressure vessel 160 from block 540 minus the radius of the nuclearreactor core 101 and the reflector thickness 211 of the reflector 140from block 520.

With the reduction in the fast neutron fluence requirement of thepressure vessel 160 and the maximum thickness of the in-vessel shield105, an in-vessel shield material on the proper Pareto front 411-413 isfound with a diffusion length able to reduce fast neutron fluencesufficiently to meet the reduction in fast neutron fluence requirementof the pressure vessel 160 while no more than the maximum thickness ofthe in-vessel shield 105. The list of factors considered in theselection of the in-vessel shield material can further include: averageand peak temperature of the nuclear reactor core 101; material andvolume of the coolant; geometry of the nuclear reactor system 100; andthe nuclear, thermal, and mechanical limits of the various in-vesselshield material to form the in-vessel shield 105.

Selecting the in-vessel shield material can include determining theeffectiveness of in-vessel shield 105 based on a computationalmulti-physics computational model that combines thermal, mechanical, andtime-dependent irradiation effects. The computational multi-physicsmodel may be a tool with time-dependent neutron fluxes. Thecomputational multi-physics model may also be used to inform thelifetime 305 and the geometric form of the in-vessel shield 105.Finishing in block 560, the in-vessel shield material selection method500 includes accepting the in-vessel shield material cost. Once anin-vessel shield material from the in-vessel shield materialoptimization chart 400 is selected that satisfies the reduction in fastneutron fluence of the pressure vessel 160 on the appropriate Paretofronts 411-413, an in-vessel shield material with the lowest cost 405 onthe appropriate Pareto front 411-413 with sufficient reduction in fastneutron fluence 406 of the pressure vessel 160 is accepted for thenuclear reactor system 100.

FIG. 6 is a flowchart depicting an in-vessel shield method 600 thatincludes in-vessel shield manufacture 601, as well as techniques forin-vessel shield design and installation 602. As shown in FIG. 6 , amanufacturing flow 601 includes blocks 610, 615, 620, and 625, which area subset of a design and installation flow 602. Manufacturing flow 601is focused on the ceramic and metallurgical processes to manufacture thein-vessel shield 105. Design and installation flow 602 includes themanufacturing flow 601 and additionally includes blocks 610 and 630.Design and installation flow 602 includes designing the in-vessel shield105 for a particular nuclear reactor system 100, which includes formingthe in-vessel shield 105 of an appropriate size and shape (e.g.,in-vessel shield tiles 131A-N) for the nuclear reactor system 100.Design and installation flow 602 also includes installing the in-vesselshield 105, e.g., by mounting in-vessel shield tiles 131A-N on theinterior wall 133 of the pressure vessel 160.

Beginning in block 605, the in-vessel shield method 600 includesselecting a nuclear reactor system 100, which includes unique parametersof the nuclear reactor system 100, such as a fast neutron fluence to thepressure vessel 160, radius of the pressure vessel 160, reflectorthickness 211, and other parameters relevant to design of the in-vesselshield 105. Nuclear reactor system 100 may include a small nuclearreactor 107, which is designed to produce up to approximately 500thermal megawatts. A smaller nuclear reactor 107 may benefit moregreatly from the space savings achieved by the in-vessel shield 105.

Moving to block 610, the in-vessel shield method 600 includes selectingtwo or more neutron absorbing materials. In addition to the materialselection method 500 of FIG. 5 , in a first example, the selecting thetwo or more neutron absorbing materials can be based on: (i) anin-vessel shield thickness 210 of the in-vessel shield 105; (ii) ageometric configuration of the in-vessel shield 105; (iii) an estimatedreduction in vessel fast fluence of the pressure vessel 160, in whichthe pressure vessel 160 encapsulates the in-vessel shield 105, and thein-vessel shield 105 encapsulates a nuclear reactor core 101; (iv) ananticipated lifetime of the in-vessel shield 105; or (v) a combinationthereof. In a second example, the selecting the two or more neutronabsorbing materials can based on: (i) an interior diameter of a pressurevessel 160; (ii) an exterior diameter of a nuclear reactor core 101;(iii) a reflector thickness 211 of a reflector 140; or (iv) acombination thereof. In a third example, the selecting the two or moreneutron absorbing materials to form the in-vessel shield 105 can bebased on: (i) a pressure vessel diameter of a pressure vessel 160 of thenuclear reactor system 100; (ii) a fast neutron fluence to the pressurevessel 160 of the nuclear reactor system 100; and (iii) neutron fluenceoutside of the nuclear reactor system 100.

The two or more neutron absorbing materials include at least one nearblack neutron absorbing material and at least one gray neutron absorbingmaterial. The selection process of block 610 is like that described inblocks 510, 520, 530, 540, 550, and 560 of the in-vessel shield materialselection method 500 of FIG. 5 . The near black neutron absorbingmaterial includes a composited ceramic material, while the gray neutronabsorbing material includes a heavy metal material. Selecting the nearblack neutron absorbing material can include selecting anisotopically-tailored near black neutron absorbing material.

Composited ceramic material can include boron carbide (B₄C), hafniumcarbide (HfC), or gadolinium oxide (Gd₂O₃); each acting as strongneutron poisons, providing the near black neutron absorbing materialpotency in reducing neutron fluence. Composited ceramic material canfurther include aluminum oxide (Al₂O₃) or silicon carbide (SiC); thesestructural materials are hardy, inert, and have small neutroncapture-cross sections: they physically support the strong neutronpoisons while not being strong neutron poisons themselves, and may becategorized as neutron moderators. The near black neutron absorbingmaterial selected may be an isotopically-tailored near black neutronabsorbing material. Particular combinations of strong neutron poisonsand structural materials for forming the composited ceramic materialinclude: boron-10 carbide (¹⁰B₄C), a boron-10 carbide and aluminum oxidecomposite (¹⁰B₄C—Al₂O₃) of 50% boron-10 carbide by weight percent, aboron-10 carbide and silicon carbide composite (¹⁰B₄C—SiC) of 50%boron-10 carbide by weight percent, or a borated stainless steel alloyof 5% boron-10 by weight percent. Boron-10 carbide and borated stainlesssteel are not composited ceramics alone, but must be paired with otherceramics in order to form a composited ceramic.

Heavy metal material of the gray neutron absorbing material includestungsten (W), iron (Fe), Nickel (Ni), or copper (Cu), for example, inthe form of atomized tungsten heavy metal. The heavy metal material,when in the form of atomized tungsten heavy metal can include atomizedtungsten heavy metal with a tungsten content greater than or equal to90% by weight percent.

Continuing to block 615, the in-vessel shield method 600 includeseutectic sintering base powders of the near black neutron absorbingmaterial to fabricate a ceramic absorbing powder. Eutectic sintering thenear black neutron absorbing material to fabricate a ceramic absorbingpowder is performed with spark plasma sintering. For example, basepowders to form the ceramic absorbing powder include an average particlesize in the range of approximately 80 nm to 50 microns. The base powdersare spark plasma sintered to form the ceramic absorbing powder. Theceramic absorbing powder includes a plurality of composited ceramicparticles with an average particle diameter greater than or equal toapproximately 80 nanometers and less than or equal to approximately 100microns. Sintering in blocks 615 and 625 can include high-vacuumdirect-current sintering. Sintering operations in some examples may besubstituted for or include an advanced 3D printing process.

In block 620, the in-vessel shield method 600 further includeskinetically mixing the gray neutron absorbing material with the ceramicabsorbing powder fabricated in block 615, creating an in-vessel shieldmixture. This ensures thorough, impurity free dispersion.

Moving to block 625, the in-vessel shield method 600 includes cold presssintering the in-vessel shield mixture into an in-vessel shield 105. Thecold press sintering the in-vessel shield mixture into the in-vesselshield 105 includes embedding the plurality of composited ceramicparticles inside a heavy metal matrix of the heavy metal material toform the in-vessel shield 105. As noted above, composited ceramicparticles include an average particle diameter greater than or equal toapproximately 80 nanometers and less than or equal to approximately 100microns.

The cold press sintering the in-vessel shield mixture into the in-vesselshield 105 further includes forming the in-vessel shield 105 as aplurality of in-vessel shield tiles 131A-N. All or a subset of theplurality of in-vessel shield tiles 131A-N are a curved polyhedron shapeor a truncated portion thereof. The plurality of in-vessel shield tiles131A-N can include a base shape with an interlocking geometry pattern.The base shape of the in-vessel shield tiles 131A-N can be selectedbased in part on: (i) tolerance towards swelling; (ii) reduction ofneutron streaming paths; or (iii) a combination thereof.

Finishing in block 630, the in-vessel shield method 600 includesmounting the in-vessel shield 105 on an interior wall 133 of a pressurevessel 160, for example, by disposing the plurality of in-vessel shieldtiles 131A-N on the interior wall 133 of the pressure vessel 160.Interior wall 133 is formed of a continuous surface or a plurality ofdiscontinuous surfaces. Hence, block 630 can include covering thecontinuous or discontinuous surfaces of the interior wall 133 with theplurality of in-vessel shield tiles 131A-N by joining the interlockinggeometry pattern.

The scope of protection is limited solely by the claims that now follow.That scope is intended and should be interpreted to be as broad as isconsistent with the ordinary meaning of the language that is used in theclaims when interpreted in light of this specification and theprosecution history that follows and to encompass all structural andfunctional equivalents. Notwithstanding, none of the claims are intendedto embrace subject matter that fails to satisfy the requirement ofSections 101, 102, or 103 of the Patent Act, nor should they beinterpreted in such a way. Any unintended embracement of such subjectmatter is hereby disclaimed.

It will be understood that the terms and expressions used herein havethe ordinary meaning as is accorded to such terms and expressions withrespect to their corresponding respective areas of inquiry and studyexcept where specific meanings have otherwise been set forth herein.Relational terms such as first and second and the like may be usedsolely to distinguish one entity or action from another withoutnecessarily requiring or implying any actual such relationship or orderbetween such entities or actions. The terms “comprises,” “comprising,”“includes,” “including,” “has,” “having,” “with,” “formed of,” or anyother variation thereof, are intended to cover a non-exclusiveinclusion, such that a process, method, article, or apparatus thatcomprises or includes a list of elements or steps does not include onlythose elements or steps but may include other elements or steps notexpressly listed or inherent to such process, method, article, orapparatus. An element preceded by “a” or “an” does not, without furtherconstraints, preclude the existence of additional identical elements inthe process, method, article, or apparatus that comprises the element.

In addition, in the foregoing Detailed Description, it can be seen thatvarious features are grouped together in various examples for thepurpose of streamlining the disclosure. This method of disclosure is notto be interpreted as reflecting an intention that the claimed examplesrequire more features than are expressly recited in each claim. Rather,as the following claims reflect, the subject matter to be protected liesin less than all features of any single disclosed example. Thus, thefollowing claims are hereby incorporated into the Detailed Description,with each claim standing on its own as a separately claimed subjectmatter.

While the foregoing has described what are considered to be the bestmode and/or other examples, it is understood that various modificationsmay be made therein and that the subject matter disclosed herein may beimplemented in various forms and examples, and that they may be appliedin numerous applications, only some of which have been described herein.It is intended by the following claims to claim any and allmodifications and variations that fall within the true scope of thepresent concepts.

1. A nuclear reactor system comprising: a pressure vessel including aninterior wall; a nuclear reactor core located within the interior wallof the pressure vessel, wherein the nuclear reactor core includes a fuelelement array of a plurality of fuel elements and at least one moderatorelement; a reflector located inside the pressure vessel that includes aplurality of reflector blocks laterally surrounding the plurality offuel elements and the at least one moderator element; and an in-vesselshield located on the interior wall of the pressure vessel to surroundthe plurality of reflector blocks, wherein: the in-vessel shield isformed of two or more neutron absorbing materials, and the two moreneutron absorbing materials include a near black neutron absorbingmaterial and a gray neutron absorbing material.
 2. The nuclear reactorsystem of claim 1, wherein: the near black neutron absorbing materialincludes a composited ceramic material; the gray neutron absorbingmaterial includes a heavy metal material; and the composited ceramicmaterial and the heavy metal material form the in-vessel shield.
 3. Thenuclear reactor system of claim 2, wherein: the composited ceramicmaterial includes boron carbide (B₄C), hafnium carbide (HfC), orgadolinium oxide (Gd₂O₃); and the heavy metal material includes tungsten(W), iron (Fe), Nickel (Ni), or copper (Cu).
 4. The nuclear reactorsystem of claim 3, wherein: the composited ceramic material furtherincludes aluminum oxide (Al₂O₃) or silicon carbide (SiC).
 5. The nuclearreactor system of claim 2, wherein: the composited ceramic materialincludes: boron-10 carbide (¹⁰B₄C), a boron-10 carbide and aluminumoxide composite (¹⁰B₄C—Al₂O₃) of 50% boron-10 carbide by weight percent,a boron-10 carbide and silicon carbide composite (¹⁰B₄C—SiC) of 50%boron-10 carbide by weight percent, or a borated stainless steel alloyof 5% boron-10 by weight percent; and the heavy metal material includesatomized tungsten heavy metal with a tungsten content greater than orequal to 90% by weight percent.
 6. The nuclear reactor system of claim2, wherein: the composited ceramic material includes a plurality ofcomposited ceramic particles with an average particle diameter greaterthan or equal to approximately 80 nanometers and less than or equal toapproximately 100 microns.
 7. The nuclear reactor system of claim 6,wherein: the plurality of composited ceramic particles are embeddedinside a heavy metal matrix of the heavy metal material to form thein-vessel shield.
 8. The nuclear reactor system of claim 7, wherein: thein-vessel shield is formed as a plurality of in-vessel shield tiles; andthe plurality of in-vessel shield tiles are disposed on the interiorwall.
 9. The nuclear reactor system of claim 8, wherein all or a subsetof the plurality of in-vessel shield tiles are a curved polyhedron shapeor a truncated portion thereof.
 10. The nuclear reactor system of claim8, wherein: the plurality of in-vessel shield tiles include a base shapewith an interlocking geometry pattern.
 11. The nuclear reactor system ofclaim 10, wherein: the interior wall is formed of a continuous surfaceor a plurality of discontinuous surfaces; and the interlocking geometrypattern of the plurality of in-vessel shield tiles are joined to coverand collectively form an in-vessel shield liner on the continuous ordiscontinuous surfaces of the interior wall.
 12. The nuclear reactorsystem of claim 11, further comprising: an inner density of activationproduct within the nuclear reactor core; and an outer density ofactivation product outside the nuclear reactor core; wherein the outerdensity of activation product is lower than the inner density ofactivation product during operation of the nuclear reactor core.
 13. Thenuclear reactor system of claim 1, wherein: a reflector block of theplurality of reflector blocks has a reflector thickness, a reflectordiffusion coefficient, and a reflector macroscopic absorption crosssection; the in-vessel shield has an in-vessel shield thickness, anin-vessel shield diffusion coefficient, and an in-vessel shieldmacroscopic absorption cross section; the reflector block and thein-vessel shield have a combined diffusion length, based at least on:the reflector thickness, the reflector diffusion coefficient, thereflector macroscopic absorption cross section, the in-vessel shieldthickness, the in-vessel shield diffusion coefficient, and the in-vesselshield macroscopic absorption cross section; and the reflector thicknessand the in-vessel shield thickness added together are less than doublethe combined diffusion length.
 14. The nuclear reactor system of claim1, wherein: each of the fuel elements includes a nuclear fuel; thenuclear fuel includes a fuel compact comprised of coated fuel particlesembedded inside a high-temperature matrix; and the high-temperaturematrix includes silicon carbide, zirconium carbide, titanium carbide,niobium carbide, tungsten, molybdenum, or a combination thereof.
 15. Thenuclear reactor system of claim 14, wherein: the coated fuel particlesincludes tristructural-isotropic (TRISO) fuel particles orbistructural-isotropic (BISO) fuel particles.
 16. The nuclear reactorsystem of claim 15, further comprising a plurality of control drums,wherein: the control drums are interspersed or disposed within thereflector.
 17. The nuclear reactor system of claim 1, wherein thenuclear reactor system includes a gas-cooled high-temperature nuclearreactor, a molten salt cooled nuclear reactor, fuel-in-salt nuclearreactor, or a sodium-cooled fast nuclear reactor.
 18. The nuclearreactor system of claim 1, further comprising a coolant that flowsthrough the plurality of fuel elements, wherein the coolant includeshelium, FLiBe molten salt formed of lithium fluoride (LiF) and berylliumfluoride (BeF₂), sodium, He, HeXe, CO₂, neon, or HeN.
 19. A methodcomprising: selecting two or more neutron absorbing materials to form anin-vessel shield, wherein the two or more neutron absorbing materialsinclude a near black neutron absorbing material and a gray neutronabsorbing material; eutectic sintering the near black neutron absorbingmaterial to fabricate a ceramic absorbing powder; kinetically mixing thegray neutron absorbing material and the ceramic absorbing powder tocreate an in-vessel shield mixture; and cold press sintering thein-vessel shield mixture into an in-vessel shield.
 20. The method ofclaim 19, wherein: the near black neutron absorbing material includes acomposited ceramic material; and the gray neutron absorbing materialincludes a heavy metal material.
 21. The method of claim 20, wherein:the composited ceramic material includes boron carbide (B₄C), hafniumcarbide (HfC), or gadolinium oxide (Gd₂O₃); and the heavy metal materialincludes tungsten (W), iron (Fe), Nickel (Ni), or copper (Cu).
 22. Themethod of claim 21, wherein: the composited ceramic material furtherincludes aluminum oxide (Al₂O₃) or silicon carbide (SiC).
 23. The methodof claim 20, wherein: the composited ceramic material includes at leastone of: boron-10 carbide (¹⁰B₄C), a boron-10 carbide and aluminum oxidecomposite (¹⁰B₄C—Al₂O₃) of 50% boron-10 carbide by weight percent, aboron-10 carbide and silicon carbide composite (¹⁰B₄C—SiC) of 50%boron-10 carbide by weight percent, or a borated stainless steel alloyof 5% boron-10 by weight percent; and the heavy metal material includesatomized tungsten heavy metal with a tungsten content greater than orequal to 90% by weight percent.
 24. The method of claim 20, wherein: thecomposited ceramic material includes a plurality of composited ceramicparticles with an average particle diameter greater than or equal toapproximately 80 nanometers and less than or equal to approximately 100microns.
 25. The method of claim 19, wherein: the ceramic absorbingpowder includes a plurality of composited ceramic particles with anaverage particle diameter greater than or equal to approximately 80nanometers and less than or equal to approximately 100 microns.
 26. Themethod of claim 24, wherein: the cold press sintering the in-vesselshield mixture into the in-vessel shield includes embedding theplurality of composed ceramic particles inside a heavy metal matrix ofthe heavy metal material to form the in-vessel shield.
 27. The method ofclaim 26, wherein the cold press sintering the in-vessel shield mixtureinto the in-vessel shield further includes forming the in-vessel shieldas a plurality of in-vessel shield tiles.
 28. The method of claim 27,further comprising: disposing the plurality of in-vessel shield tiles onan interior wall of a pressure vessel.
 29. The method of claim 28,wherein all or a subset of the plurality of in-vessel shield tiles are acurved polyhedron shape or a truncated portion thereof.
 30. The methodof claim 28, wherein: the plurality of in-vessel shield tiles include abase shape with an interlocking geometry pattern.
 31. The method ofclaim 30, wherein: the interior wall is formed of a continuous surfaceor a plurality of discontinuous surfaces; and the method furthercomprises covering the continuous or discontinuous surfaces of theinterior wall with the plurality of in-vessel shield tiles by joiningthe interlocking geometry pattern.
 32. The method of claim 30, furthercomprising selecting the base shape of the in-vessel shield tiles, basedin part on: (i) tolerance towards swelling; (ii) reduction of neutronstreaming paths; or (iii) a combination thereof.
 33. The method of claim32, wherein the base shape is a curved polyhedron shape or a truncatedportion thereof.
 34. The method of claim 19, wherein the selecting thetwo or more neutron absorbing materials is based on: (i) an in-vesselshield thickness of the in-vessel shield; (ii) a geometric configurationof the in-vessel shield; (iii) an estimated reduction in vessel fastfluence of a pressure vessel, in which the pressure vessel encapsulatesthe in-vessel shield, and the in-vessel shield encapsulates a nuclearreactor core; (iv) an anticipated lifetime of the in-vessel shield; or(v) a combination thereof.
 35. The method of claim 19, wherein theselecting the two or more neutron absorbing materials is based on: (i)an interior diameter of a pressure vessel; (ii) an exterior diameter ofa nuclear reactor core; (iii) a reflector thickness of a reflector; or(iv) a combination thereof.
 36. The method of claim 19, wherein eutecticsintering the near black neutron absorbing material to fabricate theceramic absorbing powder is performed with spark plasma sintering. 37.The method of claim 19, further comprising: selecting a reflector blockto pair with the in-vessel shield; and selecting an in-vessel shieldthickness of the in-vessel shield; wherein: the reflector block has areflector thickness, a reflector diffusion coefficient, and a reflectormacroscopic absorption cross section; the in-vessel shield has anin-vessel shield diffusion coefficient and an in-vessel shieldmacroscopic absorption cross section; the reflector block and thein-vessel shield have a combined diffusion length, based at least on:the reflector thickness, the reflector diffusion coefficient, thereflector macroscopic absorption cross section, the selected in-vesselshield thickness, the in-vessel shield diffusion coefficient, and thein-vessel shield macroscopic absorption cross section; and the reflectorthickness and the selected in-vessel shield thickness added together areless than double the combined diffusion length.
 38. The method of claim19, wherein selecting the near black neutron absorbing material includesselecting an isotopically-tailored near black neutron absorbingmaterial.
 39. The method of claim 19, further comprising: selecting anuclear reactor system; and wherein: selecting the two or more neutronabsorbing materials to form the in-vessel shield is based on: (i) apressure vessel diameter of a pressure vessel of the nuclear reactorsystem; (ii) a fast neutron fluence to the pressure vessel; and (iii)neutron fluence outside of the nuclear reactor system.
 40. The method ofclaim 19, further comprising: mounting the in-vessel shield on aninterior wall of a pressure vessel.